Similar Documents at Surry |
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Category:Drawing
MONTHYEARML20034F0532020-01-31031 January 2020 Proposed License Amendment Request: Allowed Outage Time Extension for Replacement of Reserve Station Service Transformer C 5KV Cables to Transfer Bus F Response to Request for Clarification and Additional Information ML19105B0552019-04-15015 April 2019 Superseded Pages Per Revision No. 01 to Environmental Sampling Program ML0501805082004-07-19019 July 2004 Associated System Drawings in Support of North Anna Power Station Units 1 & 2, and Surry Power Station Units 1 & 2, ASME Section XI Inservice Inspection Program End of Intervals System Pressure Testing ML0413302072004-05-0505 May 2004 Drawing 11548-FM-1E, Rev 10, Machinery Location Reactor Continment Sections A-A, X-X & F-F Surry Power Station - Unit 2. Sheet 1 of 1 ML0413302082004-05-0505 May 2004 Drawing 11548-FM-1F, Rev 9, Mach. Loc. Reactor Cont. Sections B-B. E-E. Y-Y & Z-Z Surry Power Station - Unit 2. Shet 1 of 1 ML0413302092004-05-0505 May 2004 Drawing 11548-FM-10, Rev 9, Mach Loc Reactor Cont Sections C-C & D-D Surry Power Station - Unit 2. Sheet 1 of 1 ML0414804032004-01-30030 January 2004 Drawing 11448-FE-7B, Rev 21, Wiring Diagram Annunciator (ANN-VSP) Surry Power Station - Unit 1. Sheet 1 of 1 ML0414804142002-05-29029 May 2002 Drawing 11448-FB-27A, Rev 16, Plumbing and Fire Protection Service Building Surry Power Station - Unit 1. Sheet 1 of 1 ML0414804152002-01-31031 January 2002 Drawing 11448-FB-25E, Rev 22, Ventilation & Air Conditioning Service Building - El. 9' - 6 Surry Power Station - Unit 1. Sheet 1 of 1 ML19095A2821978-02-22022 February 1978 Response to Request for Information on LOCA Pipe Break Inside Reactor Vessel Cavity ML19095A2841978-02-15015 February 1978 Presenting, Unit 2 Steam Generator Surveillance Program Outage Schedule & Information of Program ML19093B0471977-12-23023 December 1977 Request for NRC Review of Test Performance Data on Defense Apparel Air Hood & for Permission to Continue Using of Hood Indicated by VEPCO Letter of 12/6/1977 with Serial No. 558 ML19093B0821977-12-0101 December 1977 Response to a Letter of 09/14/1977 Regarding the Potential for a Boron Dilution Accident at Surry Unit Nos. 1 & 2 Due to Injection of Contents of Naoh Tank Into Reactor Coolant System ML19093A9671977-09-0101 September 1977 Letter Serial # 374 Dated 08/29/1977 Provided Information Regarding NPSH Available for Lead Safety Injection Pumps at Surry. Determined That Procedures Should Be Implemented to Throttle Valves in Lhsi Pump Discharge ML19098B4441977-06-28028 June 1977 Provide Requested Information Relative to Site Settlement to Close Previously Unresolved Item Identified as 76-15/1, Site Settlement Surveillance ML19095A5541977-01-11011 January 1977 Submit Licensee Event Report No. AO-S1-76-08 Re Broken Stud from 2A Steam Generator Girth Strap with Unit 2 ML19105A1811975-12-31031 December 1975 Submit Supplement to Initial Report Dated 12/4/1975, Which Responded to IE Bulletin 75-04B, as Requested ML19098B2971975-10-28028 October 1975 10/28/1975 Letter Supplemental Information in Technical Specification Change No. 33 2020-01-31
[Table view] Category:Letter
MONTHYEARIR 05000280/20244012024-02-0606 February 2024 Security Baseline Inspection Report 05000280/2024401, 05000281/2024401 and 07200002/2024401 IR 05000280/20230102024-01-18018 January 2024 Age Related Degradation Inspection Report 05000280-2023010 and 05000281-2023010 ML23312A1922024-01-18018 January 2024 Issuance of Amendment Nos. 316 & 316 Regarding a Risk Informed Approach for Tornado Classification of the Fuel Handling Trolley Support Structure ML24003A9022024-01-0303 January 2024 Stations, Units 1 and 2, Emergency Plan Revision - Relocation of the Technical Support Center (TSC) - Editorial Correction IR 05000280/20233012023-12-28028 December 2023 NRC Operator License Examination Report 05000280/2023301 and 05000281/2023301 ML23283A3052023-12-20020 December 2023 Review of Appendix F to DOM-NAF2, Qualification of the Framatome ORFEO-GAIA and ORFEO-NMGRID CHF Correlations in the Dominion Energy VIPRE-D Computer Code (EPID L-2022-LLT-0003) (Nonproprietary) ML23352A3692023-12-18018 December 2023 Associated Independent Spent Fuel Storage Installation, Revision to Emergency Plan Report of Change ML23334A2342023-11-30030 November 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation ML23242A2292023-11-0707 November 2023 Issuance of Amendment Nos. 315 and 315, Regarding Emergency Procedure Relocation of the Technical Support Center ML23200A2622023-11-0202 November 2023 Issuance of Amendment Nos. 314 and 314, Regarding Technical Specification and Spent Fuel Pool Criticality Change IR 05000280/20230032023-10-25025 October 2023 Integrated Inspection Report 05000280/2023003 and 05000281/2023003 ML23286A0372023-10-13013 October 2023 Annual Submittal of Technical Specifications Bases Changes Pursuant to Technical Specification 6.4.J ML23285A0922023-10-12012 October 2023 Inservice Testing Program for Pumps and Valves Sixth Interval Update and Associated Relief and Alternative Requests ML23277A1172023-10-0303 October 2023 Operator Licensing Examination Approval 05000280/2023301 and 05000281/2023301 ML23230A0502023-10-0202 October 2023 5 of the Quality Assurance Topical Report - Review of Program Changes ML23256A2672023-09-21021 September 2023 Review of the Fall 2022 Steam Generator Tube Inspection Report ML23244A2352023-09-18018 September 2023 Request to Use a Subsequent Edition of the ASME Code, Section XI ML23230A0012023-09-15015 September 2023 Request to Use a Provision of a Previous Edition of the ASME Boiler and Pressure Vessel Code, Section XI, for Non-Destructive Examinations IR 05000280/20220042023-09-0101 September 2023 Reissue - Surry Power Station - Integrated Inspection Report 05000280/2022004 and 05000281/2022004 ML23244A0142023-08-30030 August 2023 Inservice Inspection Owner'S Activity Report IR 05000280/20230052023-08-23023 August 2023 Updated Inspection Plan for Surry Power Station, Units 1 and 2 (Report 05000280/2023005 and 05000281/2023005) ML23226A1862023-08-10010 August 2023 Proposed License Amendment Request - Reclassification of Regulatory Guide 1.97 Variable for Low Head Safety Injection Flow Instrumentation ML23220A1492023-08-0707 August 2023 Independent Spent Fuel Storage Installation - Submittal of Cask Registrations for Spent Fuel Storage ML23220A1482023-08-0707 August 2023 Proposed Emergency Plan Revision - Relocation of the Technical Support Center (TSC) Response to Request for Additional Information IR 05000280/20230022023-08-0404 August 2023 Integrated Inspection Report 05000280/2023002 and 05000281/2023002 ML23208A0922023-07-26026 July 2023 Request for Approval of Appendix F of Fleet Report DOM-NAF-2-P Qualification of Framatome ORFEO-GAIA and OORFE-NMGRID CHF Correlations in the Dominion Energy Vipre-D Computer Code Response ML23207A1102023-07-26026 July 2023 NRC Regulatory Issues Summary 2023-01 Preparation and Scheduling of Operator Licensing Examinations ML23198A0322023-07-18018 July 2023 Regulatory Audit Report Regarding Review of License Amendment to Modify Emergency Plan Staff Augmentation Times ML23193A9382023-07-18018 July 2023 Closeout of Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors ML23173A0602023-07-11011 July 2023 Audit Summary for License Amendment Request Revision of the Emergency Plan to Support Relocation of the Technical Support Center ML23136B1392023-06-29029 June 2023 Issuance of Amendment No. 313 Administrative Changes to Subsequent Renewed Operating Licenses and Technical Specifications ML23178A1682023-06-26026 June 2023 2022 Annual Report of Emergency Core Cooling System (ECCS) Model Changes Pursuant to the Requirements of 10 CFR 50.46 ML23152A0432023-06-22022 June 2023 Audit Plan License Amendment Requests to Change Emergency Plan Staff Augmentation Times ML23153A1732023-06-16016 June 2023 Correction to Amendment Nos. 346 & 286 Millstone, 294 & 277 North Anna, 311 & 311 Surry, and 225 Summer to Revise Technical Specifications to Adopt TSTF-554,Rev Reactor Coolant Leakage Requirement ML23163A2352023-06-12012 June 2023 Request to Use a Later Edition of the ASME Boiler and Pressure Vessel Code, Section XI, for the SPS Sixth Inservice Inspection (Isi) Intervals and VCSNS Fifth ISI Interval ML23151A6612023-05-31031 May 2023 ES-2.1-1, Surry 2023-301 Corporate Notification Letter (Revised) ML23167B0072023-05-31031 May 2023 Proposed Emergency Plan Revision - Relocation of the Technical Support Center (TSC) Supplemental Information - LOCA Dose Calculation Summary ML23117A0952023-05-10010 May 2023 Audit Plan License Amendment Request to Support Relocation of the Technical Support Center ML23130A3742023-05-10010 May 2023 Request to Delay NRC Intial Licensing Examination 2024-02-06
[Table view] Category:Report
MONTHYEARML23167B0072023-05-31031 May 2023 Proposed Emergency Plan Revision - Relocation of the Technical Support Center (TSC) Supplemental Information - LOCA Dose Calculation Summary ML23103A2282023-04-12012 April 2023 Stations Units 1 and 2; Millstone Power Station Units 2 and 3, DOM-NAF-2-P/NP-A, Revision 0.4, Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code ML23054A1922023-02-22022 February 2023 ASME OM Code Inservice Testing Program Alternative Request V-01, Use of Mechanical Agitation Process for Pressure Isolation Valve 1-SI-241 Seat Leakage Testing Supplemental Information ML22353A6202022-12-19019 December 2022 Request for Approval of Appendix F Fleet Report DOM-NAF-2-P, Qualification of the Framatome ORFEO-GAIA and ORFEO-NMGRID CHF Correlations in the Dominion Energy VIPRE-D Computer Code ML22263A3642022-05-23023 May 2022 Alert and Notification System Evaluation Report - ANS Evaluation Redacted Pages 1 - 95 ML22263A3692022-05-0101 May 2022 Alert and Notification System Evaluation Report - ANS Evaluation Redacted Pages 96-190 ML21175A2472021-06-24024 June 2021 2020 Annual Report of Emergency Core Cooling System (ECCS) Model Changes Pursuant to the....- ML20091L9852021-05-0404 May 2021 Record of Decision ML20091L9832021-05-0404 May 2021 SLR License ML21042B3212021-02-11011 February 2021 Stations, Units 1 & 2; Millstone Power Station, Units 2 & 3 - Request for Approval of Fleet Report DOM-NAF-2 Qualification of the Framatome BWU-I CHF Correlation in the Dominion Energy VIPRE-D Computer Code ML20296A6802020-10-22022 October 2020 ASME Section XI Inservice Inspection Program Request for NRC Approval of Proposed Alternative S1-I5-ISL-05 and S2-15-ISL-06 ML20247J6162020-09-0303 September 2020 Request to Use a Provision of a Later Edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI ML20203M1602020-07-20020 July 2020 VA Elec. & Power Co., Dominion Energy Nuclear Co. Inc., Dominion Energy Sc Inc., Millstone Power Station 2, N. Anna & Surry Power Stations 1 & 2, Virgil C. Summer Station 1, Updated Anchor Darling Double Disc Gate Valve Information & Status ML20076A5762020-05-28028 May 2020 Staff Assessment of Flood Hazard Focused Evaluation and Integrated Assessment ML19249B7722019-08-29029 August 2019 Enclosure 4 - North Anna Power Station, EAL Technical Bases Document Final (Updated) ML19249B7682019-08-29029 August 2019 Enclosure 3 - Millstone Power Station EAL Technical Bases Documents Final (Updated) ML19249B7742019-08-29029 August 2019 Enclosure 5 - Surry Power Station EAL Technical Bases Document Final (Updated) ML19249B7782019-08-29029 August 2019 Enclosure 6 - Millstone Power Station, Unit 2, Comparison Matrix RCS Pot. Loss A.1 ML19235A0702019-08-19019 August 2019 Cera Blanket Website ML19235A0802019-08-19019 August 2019 Thermo-Lag Website ML19235A0742019-08-19019 August 2019 Kbs Sealbags Website ML19235A0682019-08-19019 August 2019 3M Interam Website ML19235A0722019-08-19019 August 2019 Duxseal Website ML19235A0782019-08-19019 August 2019 Mineral Wool Website ML19235A0692019-08-19019 August 2019 Bio K-10 Website ML19235A0732019-08-19019 August 2019 Gypsum Website ML19235A0712019-08-19019 August 2019 Cerafiber Website ML19235A0752019-08-19019 August 2019 Marinite Slide Plates Products - Piping Tech ML19235A0762019-08-19019 August 2019 Marinite Website ML19235A0772019-08-19019 August 2019 Micarta Website ML19235A0792019-08-19019 August 2019 Pyrocrete Website ML19204A3582019-06-30030 June 2019 WCAP-18258-NP, Revision 1, Flaw Tolerance Evaluation for Susceptible Reactor Coolant Loop Cast Austenitic Stainless Steel Elbow Components for Surry Units 1 and 2. ML19163A1542019-06-0606 June 2019 SLRA - Attachment 4D Letter - Draft to Dominion ML19140A2762019-05-20020 May 2019 Essential Fish Habitat Mapper Results for Surry Power Station ML19095A6022019-04-0202 April 2019 Enclosure 2 - SLRA Mark-Ups Change Notice 2 ML19011A1722019-01-0404 January 2019 Enclosure 3, Attachments 2C-3C - MPS3 EAL Technical Bases Document (Marked-Up) ML19011A1732019-01-0404 January 2019 Enclosure 4 - North Anna Power Station Units 1 & 2, EAL Scheme Revisions-Supporting Documents ML19011A1742019-01-0404 January 2019 Enclosure 5 - Surry Power Station, EAL Scheme Revisions-Supporting Documents ML18218A1712018-07-31031 July 2018 Attachment 5, ANP-3676NP, Revision 0, Surry Fuel-Vendor Independent Small Break LOCA Analysis, Licensing Report ML18108A1232018-04-11011 April 2018 Proposed Alternative Requests S1-15-ISl-04 and S2-15-ISl-04 Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography in Accordance with 10 CFR 50.55a(z)(1) ML17325A7792017-11-22022 November 2017 Review of Steam Generator Tube Report for the Fall 2016 Inspections During Refueling Outage 27 (CAC No. MF9683; EPID L-2017-LRO-0009) ML17339A1752017-09-0707 September 2017 Structural Integrity Associates Report No. 1601007.401, September 2017, Flaw Tolerance Evaluation of the Hot Leg Surge Line Nozzles Using ASME Code Section XI, Appendix L. ML17241A0462017-08-24024 August 2017 Submittal of Cycle 28 Startup Physics Tests Report ML17251A9002017-06-0505 June 2017 Enclosure 2, Attachment E to 17-296, WEKO-SEAL - Recommended Installation Procedures for Water and Waste-Water Piping, Miller Pipeline, Revised 03/05/17 ML16355A1722016-12-20020 December 2016 Storm Surge Technical Evaluation Report Surry Power Station Hazard Reevaluation Review ML15335A0932015-12-15015 December 2015 Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f) Seismic Hazard Reevaluations for Recommendation2.1 of the Near-Term Task Force Review ML15244B2052015-08-24024 August 2015 Virginia Electric and Power Company (Dominion) Surry Power Station Unit 1, Cycle 27 Startup Physics Tests Report ML15097A0782015-04-0707 April 2015 Soarca Surry 2005 Site File ML15072A2992015-03-16016 March 2015 Review of the Fall 2013 Steam Generator Tube Inservice Inspection Report Regarding Surry Power Station, Unit 1 Refueling Outage 25 ML15078A2882015-03-12012 March 2015 Engineering Evaluation 14-E15, Dominion Flooding Hazard Reevaluation Report, in Response to 50.54(F) Information Request Regarding Near-Term Task Force Recommendation 2.1: Flooding, Revision 1, Pp. 2-115 Through 2-218 2023-05-31
[Table view] Category:Technical
MONTHYEARML23167B0072023-05-31031 May 2023 Proposed Emergency Plan Revision - Relocation of the Technical Support Center (TSC) Supplemental Information - LOCA Dose Calculation Summary ML23103A2282023-04-12012 April 2023 Stations Units 1 and 2; Millstone Power Station Units 2 and 3, DOM-NAF-2-P/NP-A, Revision 0.4, Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code ML22263A3642022-05-23023 May 2022 Alert and Notification System Evaluation Report - ANS Evaluation Redacted Pages 1 - 95 ML22263A3692022-05-0101 May 2022 Alert and Notification System Evaluation Report - ANS Evaluation Redacted Pages 96-190 ML21175A2472021-06-24024 June 2021 2020 Annual Report of Emergency Core Cooling System (ECCS) Model Changes Pursuant to the....- ML21042B3212021-02-11011 February 2021 Stations, Units 1 & 2; Millstone Power Station, Units 2 & 3 - Request for Approval of Fleet Report DOM-NAF-2 Qualification of the Framatome BWU-I CHF Correlation in the Dominion Energy VIPRE-D Computer Code ML20296A6802020-10-22022 October 2020 ASME Section XI Inservice Inspection Program Request for NRC Approval of Proposed Alternative S1-I5-ISL-05 and S2-15-ISL-06 ML20203M1602020-07-20020 July 2020 VA Elec. & Power Co., Dominion Energy Nuclear Co. Inc., Dominion Energy Sc Inc., Millstone Power Station 2, N. Anna & Surry Power Stations 1 & 2, Virgil C. Summer Station 1, Updated Anchor Darling Double Disc Gate Valve Information & Status ML19249B7742019-08-29029 August 2019 Enclosure 5 - Surry Power Station EAL Technical Bases Document Final (Updated) ML19249B7782019-08-29029 August 2019 Enclosure 6 - Millstone Power Station, Unit 2, Comparison Matrix RCS Pot. Loss A.1 ML19249B7682019-08-29029 August 2019 Enclosure 3 - Millstone Power Station EAL Technical Bases Documents Final (Updated) ML19249B7722019-08-29029 August 2019 Enclosure 4 - North Anna Power Station, EAL Technical Bases Document Final (Updated) ML19235A0802019-08-19019 August 2019 Thermo-Lag Website ML19235A0682019-08-19019 August 2019 3M Interam Website ML19235A0692019-08-19019 August 2019 Bio K-10 Website ML19235A0702019-08-19019 August 2019 Cera Blanket Website ML19235A0792019-08-19019 August 2019 Pyrocrete Website ML19235A0712019-08-19019 August 2019 Cerafiber Website ML19235A0722019-08-19019 August 2019 Duxseal Website ML19235A0732019-08-19019 August 2019 Gypsum Website ML19235A0742019-08-19019 August 2019 Kbs Sealbags Website ML19235A0752019-08-19019 August 2019 Marinite Slide Plates Products - Piping Tech ML19235A0762019-08-19019 August 2019 Marinite Website ML19235A0772019-08-19019 August 2019 Micarta Website ML19235A0782019-08-19019 August 2019 Mineral Wool Website ML19204A3582019-06-30030 June 2019 WCAP-18258-NP, Revision 1, Flaw Tolerance Evaluation for Susceptible Reactor Coolant Loop Cast Austenitic Stainless Steel Elbow Components for Surry Units 1 and 2. ML19095A6022019-04-0202 April 2019 Enclosure 2 - SLRA Mark-Ups Change Notice 2 ML19011A1742019-01-0404 January 2019 Enclosure 5 - Surry Power Station, EAL Scheme Revisions-Supporting Documents ML19011A1732019-01-0404 January 2019 Enclosure 4 - North Anna Power Station Units 1 & 2, EAL Scheme Revisions-Supporting Documents ML19011A1722019-01-0404 January 2019 Enclosure 3, Attachments 2C-3C - MPS3 EAL Technical Bases Document (Marked-Up) ML18218A1712018-07-31031 July 2018 Attachment 5, ANP-3676NP, Revision 0, Surry Fuel-Vendor Independent Small Break LOCA Analysis, Licensing Report ML17339A1752017-09-0707 September 2017 Structural Integrity Associates Report No. 1601007.401, September 2017, Flaw Tolerance Evaluation of the Hot Leg Surge Line Nozzles Using ASME Code Section XI, Appendix L. ML17241A0462017-08-24024 August 2017 Submittal of Cycle 28 Startup Physics Tests Report ML17251A9002017-06-0505 June 2017 Enclosure 2, Attachment E to 17-296, WEKO-SEAL - Recommended Installation Procedures for Water and Waste-Water Piping, Miller Pipeline, Revised 03/05/17 ML16355A1722016-12-20020 December 2016 Storm Surge Technical Evaluation Report Surry Power Station Hazard Reevaluation Review ML15244B2052015-08-24024 August 2015 Virginia Electric and Power Company (Dominion) Surry Power Station Unit 1, Cycle 27 Startup Physics Tests Report ML14086A1732014-04-0909 April 2014 Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-Ichi Nuclear Power Plant Accident ML14079A0182014-02-27027 February 2014 30-Day Report of Emergency Core Cooling System Model Changes Pursuant to the Requirements of 10 CFR 50.46 ML14002A1452014-02-19019 February 2014 Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies) ML14043A4442014-02-12012 February 2014 Mega-Tech Services, LLC Technical Evaluation Report Regarding the Overall Integrated Plan for Surry Power Station, Units 1 and 2, TAC Nos.: MF1002 and MF1003 (Revision 2) ML13179A0152013-05-31031 May 2013 Units I and 2, Qualification of the ABB-NV and Wlop CHF Correlations in the Dominion VIPRE-D Computer Code ML11215A0592011-06-30030 June 2011 WCAP-17345-NP, Rev 2, H*: Resolution of NRC Technical Issue Regarding Tubesheet Bore Eccentricity (3-Loop Model 44F/ Model 51F), Attachment 5 to 11-403 ML1023904192010-08-31031 August 2010 DOM-NAF-2, Rev. 0.2-A, Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code. ML1026002162010-05-13013 May 2010 Engineering Transmittal, ET-CEP-10-0006, Revision 0, Evaluation of Aluminum Conduit Seal Penetration Fire Tests, Dated May 13, 2010, North Anna and Surry Power Stations Units 1 and 2 (Redacted Version), Enclosure 2 ML0930701642009-10-15015 October 2009 Technical Report No. EE-0116, Revision 5 ML0926606162009-08-27027 August 2009 LTR-SGMP-09-108 NP-Attachment, Response to NRC Request for, Additional Information on H*; Model 44Fand Model 51F Steam Generators. ML0921504622009-06-30030 June 2009 WCAP-17092-NP, Revision 0, H*: Alternate Repair Criteria for the Tubesheet Expansion Region in Steam Generators with Hydraulically Expanded Tubes (Model 51F), Attachment 6 ML0921202022009-05-0606 May 2009 LTR-SGMP-09-64 Rev. 2 Np, Surry, Unit 1, Support for Emergency Technical Specification Amendment to Address Permeability Variation Signals at the Tube Ends. ML0826107152008-09-17017 September 2008 Final Precursor Analysis - Surry 2 (LER 281/06-002) ML0823200152008-08-18018 August 2008 Cycle 22 Startup Physics Tests Report 2023-05-31
[Table view] |
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e VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND,VIRGINIA 23261 December 1, 197].
Mr. Edson G. Case Serial No. 424/091477 Acting Director of Nuclear Reactor Regulation PO&M/DLB:dgt United States Nuclear Regulatory Commission Docket Nos. 50-280 Washington, .D. C. 20555 50-281 Attention: Mr. Robert W. Reid, Chief License Nos. DPR-32 Operating Reactors Branch 4 DPR-37
Dear Sir:
This is in response to your letter of September 14, 1977, regarding the potential for a boron dilution accident at Surry Unit Nos. 1 and 2 due to the injection of the contents of the NaOH tank into the reactor coolant system.
In the case cited in your letter, it was concluded that for certain conserva-tive assumptions, the injection of the NaOH contents into the reactor coolant due to the malposition of a single isolation valve could result in reactor criticality with the control rods inserted.
The attached analysis has been completed to determine the potential for this accident at our Surry Unit Nos. 1 and.2. Analysis revealed that multiple
- component failures and/or operation errors combined with unrealistic initial conditions would be required to initiate this accident. In addition, the con-sequences of this accident were found to be acceptable. Based on this analysis, it is concluded that the postulated accident is neither serious nor credible.
No further analysis and no corrective action is considered necessary.
I
~ r ' tr~l yours, I
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/) u/1 yftfW) Ul '/:_~<-,;/
J Vice resident-Power pp)t, an~ Production Operations Attachment
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ANALYSIS OF POTENTIAL FOR BORON DILUTION OF THE REACTOR COOLANT SYSTEM BY DRAINAGE OF THE CHEMICAL ADDITION TANK Introduction The purpose of this analysis is to investigate the potential for the boron dilution of the reactor coolant by drainage of the NaOH tank into the reactor coolant system. This is in response to an NRC letter dated September 14, 1977 which cited a dilution incident at another operating PWR facility, and requested this analysis for Surry Power Station, unit ~os. 1 and 2.
System Description
Reactor Coolant System, Residual Heat Removal System, and Containment Spray System components and piping relevant to this analysis are shown in Figure 1.
The Refueling Water Storage Tank (RWST) normally contains not less 350,000 gallons of borated water at not less than 2000 ppm. The RWST provides borated water to the safety injection system and the containment spray system under accident conditions. In addition, the RWST provides borated water to fill the reactor cavity during refueling operations. The RWST water temperature is main-tained at approximately 40 degrees F by circulation through a cooling loop which contains two recirculation pumps, two chilled water cool~rs and two mechanical refrigeration units. The chemical addition tank (CAT) contains n_ot less than 3360 gallons of solution with a NaOH concentration of not less than 18 percent by weight. The CAT provides a balanced gravity feed of NaOH solution for iodine removal during containment spray operation.
The Residual Heat Removal (RHR) system provides cooldown below approximately 350 degrees F and continuing removal of decay heat during unit shutdown. This system includes two pumps, two heat exchangers, and motor-operated valves for isolation from the reactor coolant system. A reactor cavity dewatering line (RH-20) runs from the discharge of the RHR pumps to the discharge of the refuel-ing water recirculation pumps. ~his line, RH-20, is used to return water from the reactor cavity to the RWST following completion of refueling.
System Operations Under normal operation the valve lineup is as shown in Figure 1.
Suction to the refueling water circulating pumps is normally from the upper tap through valve 1-CS-26. The lower tap through valve 1-CS-27 is only opened when higher flow rates through the cooling loop are necessary for mixing prior to chemistry sampling or for accelerated cooldown as might be required prior to unit startup.
The chemical addition tank outlet valves, MOV-CS-102A and Hov~cs-102B, are normally closed. They are opened only on containment spray initiation and once per month for valve testing. Testing of .each involves closing of the upstream and downs trea,m isolation valves, opening and closing the MOV; and reopening the isolation valves. The valves are tested one at a time to ensure that a path for
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NaOH addition is always available.
Under normal operations, the RHR system is shutdown and isolated from the reactor coolant system with valves MOV-RHR-1700, 1701, 1720A and 1720B closed.
During shutdown, RHR is in service with these valves open, 1 pump and 1 or 2 heat exchangers in service. Valve HOV-RHR-100 is shut during normal operation and during normal shutdown. This valve is opened only for a short period during refueling when the reactor cavity water is returned to the RWST.
Reactor Core Conditons - Shutdown Margins During refueling shutdown, a shutdown margin of at least 10;~ reactivity is required. During cold shutdown a shutdown margin of at least 1% reactivity is required. Administrative limits require a 2% shutdown margin for cold shut-down. Under the worst conditions, i.e. beginning of core life, shutdown banks out, using our administrative limit of 2%, a boron concentration of not greater than 1400 ppm is required. This value of 1400 ppm represents the maximum boron concentration which would be required to maintain a 2% shutdown condition at any time in core life, with the shutdown banks out.
Dilution Accident Considerations The following combination of event*s would be necessary to".'1ead to a dilution of the reactor coolant system by drainage of the chemical addition tank.
- 1) Malposition of either of the chemical addition tank discharge valves, MOV-CS-102A or 1021
- 2) A difference in level between the CAT and the RWST, i.e. RWST must be below normal level
- 3) Valve l-CS-27 must be open
- 4) The RWST recirculation loop must be in normal lineup
- 5) Valve MOV-RHR-100 must be open
- 6) Both RHR pumps must be shutdown during the above events, after which one or both pumps must be restarted to carry the dilution into the reactor coolant system This series of events would allow flow from the CAT to the RWST, dilution of the RWST, flow from the RWST through valve l-CS-27 into the recirculation loop, to the RHR *system through MOV-RHR-100 and then into the reactor coolant system.
Analysis of Events Following is an analysis of the potential for each of the events listed above.
Opening of CAT Discharge Valves Malpositioning of either valve HOV-CS-102A or MOV-CS-102B is necessary to
l e e establish a flowpath from the CAT to the RWST. In the case cited in the NRG letter, the corresponding valve was malpositioned during a valve cycling test.
Our procedure for this test requires the closing of both the upstream and down-stream isolation valves prior to cycling the NaOH tank discharge valve. Thus,
- a flow path to the RWST would not be established without a violation of the test procedure. Accidental opening of these valves is considered extremely unlikely due to their location on the safeguards area of the control board.
Flow to the RWST If a flowpath to the RWST is established, the rate of flow and the amount discharged to the RWST would depend on the relative levels of the RWST and the chemical addition tank. Under normal conditions, the RWST would be 100% full.
If either NaOH tank discharge valve opened under this condition, there would be only a slight flow of NaOH into the RWST, this being due to the higher speci-fic gravity of the NaOH solution. Ultimately, due to specific gravity difference, the NaOH would settle to the bottom of the RWST. Assuming complete stratification occurred, the NaOH would settle to the bottom 5 inches of the RWST, 15 inches be-low the lowest outlet to the recirculation loop. Thus, the n~lpositioning of either NaOH tank discharge valve with a 100% level in the RWST does not present a dilution problem.
The assumption of an RWST level other than 100% presents -certain conflicts with other assumptions specified in the NRG analysis request. The only operation during which RWST level is reduced below 100% is for .filling the reactor cavity during refueling. This is inconsistent with the assumpd,on that the reactor vessel is drained to the nozzles as specified in the NRG request. There is no normal operating condition where the RWST level is below 100% and the reactor vessel is drained to the nozzles. However, while this is not realistic, it is conservative, and simplifies analysis.
The worst initial level in the RWST for this accident is in the range of 15" to 20 11
- The outlet to l-CS-27 is located at the 20 11 level. RWST levels above 20" are not specifically addressed because to have a level above 20", l-CS-27 must have been closed.
- The only other outlet to the recirculation loop .is then through l-CS-26 which is almost at the lb0% level. As discussed previously, opening of the NaOH tank discharge valve with a very high level in the RWST does not present a dilution problem.
Figure 2 summarizes calculations of the release from the RWST resulting from drainage of the CAT for RWST levels from 15i; to 20". At 20 11 of level 3670 gallons with a boron concentration of 1588 ppm are released. For levels below 20 11 , the dilution of the RWST water is greater. At 15" or below, no re-lease occurs because the water added is not sufficient to reach the outlet at 20 11
- Note that this calculation assumes complete mixing which is conservative.
Settling of the heavier NaOH solution would reduce the amount which overflows through l-CS-27. Since the water in the RWST begins overflowing to the recircu-lation loop as soon as drainage begins and much of the NaOH solution is still in the CAT as overflow continues, the actual boron concentration leaving the RWST will be significantly higher than calculated.
Flow from RWST to RHR System In order for the borated water to leave the RWST and flow into the reactor
RWST . F 1GURE 2
_ ~ - OVERFLOW io RWST
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e coolant system via the RHR system it is necessary to have the RHR system lined up for operation, with both RHR pumps shutdown and MOV-RHR-100 open.
In the shutdown condition, one RHR pump is run at all times. If HOV-RHR-100 is opened with one or both RHR pumps running, water would flow to the RWST from the. RHR system.
Valve MOV-RHR-100 is only opened during refueling for dewatering the refuel-ing cavity, at which time one RHR pump would be running. Failure of a running RHR pump would quickly cause an RHR low flow alarm, at which time the operator would start the other pump.
It is therefore unlikely that MOV-RHR-100 would ever be open at the same time that the RHR system was aligned for operation with both RHR pumps shutdown.
Reactivity Effects As explained above, the quantity and boron concentration of water leaving the RWST varies depending on RWST level. Figure 2 summarizes this data. Also in Figure 2 is a tabulation of the ultimate effect of this water reaching the primary. This calculation assumes an initial concentration of 1400 ppm in the primary. As discussed previously, 1400 ppm is the maximum boron concentration which would be required to maintain a 2% shutdown condition at,,..*any time in core life. Note that in every case a negative reactivity addition is the result. For*
lesser initial reactor coolant boron concentrations, a greater negative reactivity addition would result. For boron concentrations above approximately 1480, this accident would result in a minor net dilution (for example, at 2000 ppm, a dilution of 50 ppm would occur). However, this could never reduce the boron concentration below 1480 ppm which is well above the concentration required for 21o shutdown margin. Therefore, the consequences of this accident are not serious.
Conclusion The sequence of events necessary to cause a reactor coolant boron dilution by drainage of the sodium hydroxide tank were listed on page 2. Subsequently, the potential for each event was considered. In addition, the consequences of the accident were analyzed and determined to be acceptable.*
The NRC letter addressed a situation where a single malpositioned valve could result in a dilution. This is clearly not the case at Surry Unit Nos. 1 and 2.
At Surry this accident would require the malpositioning of at least 2 valves, the failure of at least one pump, combined with a series of plant conditions which, while possible, do not occur simultaneously under any known operational mode or evolution.
It is therefore concluded that this accident is neither serious nor credible.
No further analysis and no corrective action is considered necessary.